05000272/LER-2007-001, Re ESF Actuation of Auxiliary Feedwater Pumps in Mode 3
| ML071500219 | |
| Person / Time | |
|---|---|
| Site: | Salem  (DPR-070) |
| Issue date: | 05/23/2007 |
| From: | Joyce T Public Service Enterprise Group |
| To: | Document Control Desk, Plant Licensing Branch III-2 |
| References | |
| LR-N07-0111 LER 07-001-00 | |
| Download: ML071500219 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2722007001R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236
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r 0 PSEG Nuclear LLC MAY 1 3 2007 1 OCFR50.73 LR-N07-0111 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 LER 272/07-001 Salem Nuclear Generating Station Unit 1 Facility Operating License No. DPR-70 NRC Docket No. 50-272
SUBJECT:
ESF Actuation of Auxiliary Feedwater Pumps in Mode 3 This Licensee Event Report, "ESF Actuation of Auxiliary Feedwater Pumps in Mode 3" is being submitted pursuant to the requirements of the Code of Federal Regulations 1 OCFR50.73(a)(2)(iv)(B).
The attached LER contains no commitments. Should you have any questions or comments regarding this submittal, please contact Mr. Howard Berrick at 856-339-1862.
Sincerely, Thomas P. Joyce Site Vice President Salem Generating Station Attachments (1) 95-2168 REV. 7/99
MAY, 2 3 207 Document Control Desk Page 2 LR-N07- 0111 cc Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 [6-2oo4)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the
- 3. PAGE Salem Generating Station - Unit 11 05000272 1 OF 4
- 4. TITLE ESF Actuation of Auxiliary Feedwater Pumps in Mode 3.
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEUNIATE FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO MONTH DAY YEAR 23 FACILITY NAME DOCKET NUMBER 03 27 2007 2007 001 00 05 2
2007
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
[1 20.2203(a)(3)(i) 0l 50.73(a)(2)(i)(C) 0l 50.73(a)(2)(vii) 3 0l 20.2201(d)
[I 20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1) 0l 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B) 0_ 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
C1 50.73(a)(2)(iii)
C1 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
Dl 50.36(c)(1)(ii)(A)
E 50.73(a)(2)(iv)(A) 0l 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
Dl 50.73(a)(2)(v)(A)
El 73.71 (a)(4) 0%
D 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B) 0l 73.71 (a)(5) l[ 20.2203(a)(2)(v) 0l 50.73(a)(2)(i)(A)
Dl 50.73(a)(2)(v)(C)
Dl OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below C..0...- I-A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Howard Berrick, Senior Licensing Engineer 1856-339-1862MAU EPRA lE
- - MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FMANU" REPORTABLE
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX FCUE OEI A
AF-N
- 14. SUPPLEMENTAL REPORT EXPECTED
- 16. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On March 27, 2007 at approximately 2050, Salem Unit 1 was in Mode 3 following a planned manual reactor trip to begin a scheduled refueling outage. Operators had established initial Reactor Coolant System cooldown using the steam dumps for heat removal and the 11, 12, and 13 Auxiliary Feedwater (AFW) pumps for steam generator make-up. Steam Generator levels were being maintained lower than normal in preparation for an AFW full flow test. As plant cooldown proceeded, narrow range levels in 11, 12, and 13 Steam Generators reached the low steam generator setpoint trip, resulting in a valid ESF actuation (i.e., start signal to the AFW pumps); however all AFW pumps were already inservice. The actuation signal also generated a reactor trip signal; however, the plant was already in a shutdown condition with the reactor trip breakers open. The lowest level during this transient occurred in 13 Steam Generator and was 11.3% narrow range level. Steam generator water level was restored to a normal value and the RCS cooldown recommenced. There were no equipment failures that contributed to this event.
The cause of this event is attributed the failure of the operating crew to establish clear termination criteria for stopping the cooldown based on low S/G levels and the lack of clear termination criteria guidance in the procedure for maintaining S/G levels during a cooldown. This event is reportable in accordance with 10CFR50.73 (a)(2)(iv)(A), "any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section."
NRC F*RM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REVISION YEAR NUMBER NUMBER Salem Generating Station Unit 1 05000272 2007
- - 0 0 1-00 2 OF 4
- 17. NARRATIVE (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A)
SAFETY CONSEQUENCES AND IMPLICATIONS
There was no actual safety consequences associated with this event; sufficient cooling was always maintained. The low level in the 11, 12 and 13 S/G occurred as a result of human error and was not caused by equipment malfunction. The safety systems responded to the low S/G level as designed.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, did not occur. There was no condition that alone could have prevented the fulfillment of a safety function of a system needed to remove residual heat.
CORRECTIVE ACTIONS
- 1. Initiated a prompt investigation of event, and subsequent root cause evaluation.
- 2. Personnel involved with this event have been held accountable in accordance with PSEG policies.
- 3. Operations conducted 'stand downs' to reemphasize the application of Operations fundamentals, particularly Control Board awareness and Teamwork.
- 4. Prior to Mode ascension at the end of 1 R1 8, the following actions were taken:
a. Crew composition adjustments made, b. Repaired steam dump valve 13TB10, c. Additional oversight was assigned during critical evolutions, requiring Operations Manager presence at pre-job briefs to reinforce expectations, and d. Just-In-Time (JIT) Training on AFW control at low power.
- 5. Procedure IOP-6, Hot Standby to Cold Shutdown, will be revised to include a level band on all S/Gs prior to initiating cooldown and establishing cooldown termination criteria with respect to minimum S/G levels.
- 6. Additional corrective actions include:
a. Reinforcement of operating and human performance fundamentals.
b. An extent of condition review to identify other procedures that could render a similar outcome.
c. Presentation of event to Operations Curriculum Review Committee for inclusion into the Training program.
COMMITMENTS
No commitments are made in this LER.